• 상하수도학회지
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• 제21권1호
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• pp.83-89
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• 2007

A biological anoxic-aerobic reactor filled with porous media was operated in lab scale for the advanced wastewater treatment. The experiments were conducted for 6 months with three HRTs (4, 6, 8hr) and temperature of $23{\sim}25^{\circ}C$. Some other experimental conditions were as follows; nitrification reactor (MLSS 4,500mg/L, DO 3.3mg/L, $23{\sim}28^{\circ}C$), denitrification reactor(MLSS 8,000mg/L, ORP -100mV, Temp.$19{\sim}23^{\circ}C$). Average removal efficiencies of SS, $BOD_5$, $COD_{Cr}$, T-N, and T-P were 97.8%, 95.5%, 94.5%, 80.2%, and 60.6%, respectively. The reactor filled with porosity media showed stable removal capacity for organics and nutrients. Fast and complete nitrification and denitrification were accomplished. Maintaining high MLSS with porous media in the nitrification and denitrification reactor appears to enhance the nitrogen removal process. For the higher T-P removal, some coagulant addition process will be needed.

• 한국유체기계학회 논문집
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• 제13권5호
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• pp.5-10
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• 2010

5MW, open pool type research reactor, is commonly used to education and experimental purpose. It is necessary to prepare a standardization of system designs for considering a demand. HANARO has prepared the standardization of 5MW research reactor system designs based on the design, installation, commissioning and operating experiences of HANARO. For maintaining an open pool type reactor safety, a primary cooling system (after below, PCS) should remove the heat generated by the reactor under a reactor normal operation condition and a reactor shutdown condition. For removing the heat generated by the reactor, the PCS should maintain a required coolant flow rate. For a verification of the required flow rate, a flow network analysis of the PCS was carried under a normal operating condition. Based on the flow network analysis result, this paper describes the PCS flow characteristics of a 5MW open pool type research reactor. Through the result, it was confirmed that the PCS met design requirements including design flow rate without cavitation.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4465-4471
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• 2023

Hands-on education and training is a key part of fixing and developing technology knowledge and is an inherent part of many engineering and scientific curricula. However, access to large complex training facilities, such as nuclear reactor, could be limited by various factors, such as unavailability of those facilities in the region, high traveling costs or harmonization of the schedules of hands-on E&T with theoretical lectures and with the operational schedule of the facility. To handle the issue, several success stories have been reached with the introduction of the Internet Reactor Labs (IRL). The Internet Reactor Labs can strongly contribute to accessibility of training at research reactors and can contribute to improvements in their utilization. The paper describes the development of the Internet Reactor Lab at the VR-1 reactor of the Czech Technical University in Prague. Contrary to single-purpose IRLs, it presents various modalities of online teaching and training in experimental reactor physics and reactor operation in general as well as outreach activities that have been developed in recent years.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.357-373
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• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3549-3558
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• 2023

The jet pump can be used in a test device of a nuclear reactor for high flow amplification as it reduces inlet flow requirement and thereby size of the process components. In the present work, a miniature jet pump was designed to meet high flow amplification greater than 3. Subsequently, experiments were carried out using a test setup for design validation and performance evaluation of the jet pump for different parameters. It was observed that a minimum pressure of 0.6 bar (g) was required for the secondary fluid inside the jet pump to ensure cavitation free performance at high amplification. Spacing between the nozzle tip and the mixing chamber entry point had significant effect on the performance of the jet pump. Variation in primary flow, temperature and area ratio also affected the performance. It was observed that at high flow amplification, the analytical solution differed significantly from experimental results due to very large velocities encountered in the miniature size jet pump.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1037-1051
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• 2024

This paper presents the construction results and design of the UNIST Reactor Innovation platform for small modular reactors as a versatile testbed for exploring innovative technologies. The platform uses simulant fluids to simulate the thermal-hydraulic behavior of a reference small modular reactor design, allowing for cost-effective design modifications. Scaling analysis results for single and two-phase natural circulation flows are outlined based on the three-level scaling methodology. The platform's capability to simulate natural circulation behavior was validated through performance calculations using the 1-D system thermal-hydraulic code-based calculation. The strategies for evaluating cutting-edge technologies, such as the integration of a solid oxide electrolysis cell for hydrogen production into a small modular reactor, are presented. To overcome experimental limitations, the hardware-in-the-loop technique is proposed as an alternative, enabling real-time simulation of physical phenomena that cannot be implemented within the experimental facility's hardware. Overall, the proposed versatile innovation platform is expected to provide valuable insights for advancing research in the field of small modular reactors and nuclear-based hydrogen production.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1002-1012
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• 2024

To remove insoluble fission products, which could possibly cause reactor instability and significantly reduce heat transfer efficiency from primary system of molten salt reactor, a helium bubbling method is employed into a passive molten salt fast reactor. In this regard, two-phase flow behavior of molten salt and helium bubbles was investigated experimentally because the helium bubbles highly affect the circulation performance of working fluid owing to an additional drag force. As the helium flow rate is controlled, the change of key thermal-hydraulic parameters was analyzed through a two-phase experiment. Simultaneously, to assess the applicability of numerical model for the analysis of two-phase flow behavior, the numerical calculation was performed using the OpenFOAM 9.0 code. The accuracy of the numerical analysis code was evaluated by comparing it with the experimental data. Generally, numerical results showed a good agreement with the experiment. However, at the high helium injection rates, the prediction capability for void fraction of helium bubbles was relatively low. This study suggests that the multiphaseEulerFoam solver in OpenFOAM code is effective for predicting the helium bubbling but there exists a room for further improvement by incorporating the appropriate drag flux model and the population balance equation.